This invention relates broadly to an improvement in nuclear fuel elements for use in the core of nuclear fission reactors, and more particularly to an improved nuclear fuel element having a composite cladding container having a substrate and a metal layer on the inside surface of the substrate.
Nuclear reactors are presently being designed, constructed and operated in which the nuclear fuel is contained in fuel elements which can have various geometric shapes, such as plates, tubes, or rods. The fuel material is usually enclosed in a corrosion-resistant, non-reactive, heat conductive container or cladding. The elements are assembled together in a lattice at fixed distances from each other in a coolant flow channel or region forming a fuel assembly, and sufficient fuel assemblies are combined to form the nuclear fission chain reacting assembly or reactor core capable of a self-sustained fission reaction. The core in turn is enclosed within a reactor vessel through which a coolant is passed.
The cladding serves several purposes and two primary purposes are: first, to prevent contact and chemical reactions between the nuclear fuel and the coolant or the moderator if a moderator is present, or both if both the coolant and the moderator are present; and second, to prevent the radioactive fission products, some of which are gases, from being released from the fuel into the coolant or the moderator or both if both the coolant and the moderator are present. Common cladding materials are stainless steel, aluminum and its alloys, zirconium and its alloys, niobium (columbium), certain magnesium alloys, and others. The failure of the cladding, i.e., a loss of the leak tightness, can contaminate the coolant or moderator and the associated systems with radioactive long-lived products to a degree which interferes with plant operation.
Problems have been encountered in the manufacture and in the operation of nuclear fuel elements which employ certain metals and alloys as the clad material due to mechanical or chemical reactions of these cladding materials under certain circumstances. Zirconium and its alloys, under normal circumstances, are excellent nuclear fuel claddings since they have low neutron absorption cross sections and at temperatures below about 750.degree. F. (about 398.degree. C.) are strong, ductile, extremely stable and non-reactive in the presence of demineralized water or steam which are commonly used as reactor coolants and moderators.
However, fuel element performance has revealed a problem with the brittle splitting of the cladding due to the combined interactions between the nuclear fuel, the cladding and the fission products produced during nuclear fission reactions. It has been discovered that this undesirable performance is promoted by localized mechanical stresses due to fuel cladding differential expansion (stresses in the cladding are localized at cracks in the nuclear fuel). Corrosive fission products are released from the nuclear fuel and are present at the intersection of the fuel cracks with the cladding surface. Fission products are created in the nuclear fuel during the fission chain reaction during operation of a nuclear reactor. The localized stress is exaggerated by high friction between the fuel and the cladding.
Within the confines of a sealed fuel element, hydrogen gas can be generated by the slow reaction between the cladding, and the residual water inside the cladding may build up to levels which under certain conditions can result in localized hydriding of the cladding with concurrent local deterioration in the mechanical properties of the cladding. The cladding is also adversely affected by such gases as oxygen, nitrogen, carbon monoxide and carbon dioxide over a wide range of temperatures.
The zirconium cladding of a nuclear fuel element is exposed to one or more of the gases listed above and fission products during irradiation in a nuclear reactor and this occurs in spite of the fact that these gases and fission product elements may not be present in the reactor coolant or moderator, and further may have been excluded as far as possible from the ambient atmosphere during manufacture of the cladding and the fuel element. Sintered refractory and ceramic compositions, such as uranium dioxide and other compositions used as nuclear fuel, release measurable quantities of the aforementioned gases and fission products upon heating, such as during fuel element manufacture and further release fission products during irradiation. Particulate refractory and ceramic compositions, such as uranium dioxide powder and other powders used as nuclear fuel, have been known to release even larger quantities of the aforementioned gases during irradiation. These released gases are capable of reacting with the zirconium cladding containing the nuclear fuel.
Thus in light of the foregoing, it has been found desirable to minimize attack of the cladding from water, water vapor and other gases, especially hydrogen, reactive with the cladding from inside the fuel element throughout the time the fuel element is used in the operation of nuclear power plants. One such approach has been to find materials which will chemically react rapidly with the water, water vapor and other gases to eliminate these from the interior of the cladding, and such materials are called getters.
Another approach has been to coat the nuclear fuel material with a ceramic to prevent moisture coming in contact with the nuclear fuel material as disclosed in U.S. Pat. No. 3,108,936. U.S. Pat. No. 3,085,059 presents a fuel element including a metal casing containing one or more pellets of fissionable ceramic material and a layer of vitreous material bonded to the ceramic pellets so that the layer is between the casing and the nuclear fuel to assure uniformly good heat conduction from the pellets to the casing. U.S. Pat. No. 2,873,238 presents jacketed fissionable slugs of uranium canned in a metal case in which the protective jackets or coverings for the slugs are a zinc-aluminum bonding layer. U.S. Pat. No. 2,849,387 discloses a jacketed fissionable body comprising a plurality of open-ended jacketed body sections of nuclear fuel which have been dipped into a molten bath of a bonding material giving an effective thermally conductive bond between the uranium body sections and the container (or cladding). The coating is disclosed as any metal alloy having good thermal conduction properties with examples including aluminum-silicon and zinc-aluminum alloys. Japanese Patent Publication No. SHO 47-46559 dated Nov. 24, 1972, discloses consolidating discrete nuclear fuel particles into a carbon-containing matrix fuel composite by coating the fuel particles with a high density, smooth carbon-containing coating around the pellets. Still another coating disclosure is Japanese Patent Publication No. SHO 47-14200 in which the coating of one of two groups of pellets is with a layer of silicon carbide and the other group is coated with a layer of pyrocarbon or metal carbide.
The coating of a nuclear fuel material introduces reliability problems in that achieving uniform coatings free of faults is difficult. Further, the deterioration of the coating can introduce problems with the long-lived performance of the nuclear fuel material.
U.S. patent application Ser. No. 330,152 filed Feb. 6, 1973 discloses a method for preventing corrosion of nuclear fuel cladding consisting of the addition of a metal such as niobium to the fuel. The additive can be in the form of a powder, provided the subsequent fuel processing operation does not oxidize the metal, or incorporated into the fuel element as wires, sheets or other forms in, around, or between fuel pellets.
Document GEAP-4555 dated Feb. 1964 discloses a composite cladding of a zirconium alloy with an inner lining of stainless steel metallurgically bonded to the zirconium alloy, and the composite cladding is fabricated by use of extrusion of a hollow billet of the zirconium alloy having an inner lining of stainless steel. This cladding has the disadvantage that the stainless steel develops brittle phases, and the stainless steel layer involves a neutron absorption penalty of about ten to fifteen times the penalty for a zirconium alloy layer of the same thickness.
U.S. Pat. No. 3,502,549 discloses a method of protecting zirconium and its alloys by the electrolytic deposition of chrome to provide a composite material useful for nuclear reactors. A method for electrolytic deposition of copper on Zircaloy - 2 surfaces and subsequent heat treatment for the purpose of obtaining surface diffusion of the electrolytically deposited metal is presented in Energia Nucleare Volume 11, number 9 (Sept. 1964) at pages 505-508. In Stability and Compatibility of Hydrogen Barriers Applied to Zirconium Alloys by F. Brossa et al. (European Atomic Energy Community, Joint Nuclear Research Center, EUR 4098e 1969), methods of deposition of different coatings and their efficiency as hydrogen diffusion barriers are described along with an Al--Si coating as the most promising barrier against hydrogen diffusion. Methods for electroplating nickel on zirconium and zirconium tin alloys and heat treating these alloys to produce alloy-diffusion bonds are disclosed in Electroplating on Zirconium and Zirconium-Tin by W. C. Schickner et al. (BMl-757, Technical Information Service, 1952). U.S. Pat. No. 3,625,821 presents a fuel element for a nuclear reactor having a fuel cladding tube with the inner surface of the tube being coated with a retaining metal of low neutron capture cross section such as nickel and having finely dispersed particles of a burnable poison disposed therein. Reactor Development Program Progress Report of Aug., 1973 (ANL-RDP-19) discloses a chemical getter arrangement of a sacrificial layer of chromium on the inner surface of a stainless steel cladding.
Another approach has been to introduce a barrier between the nuclear fuel material and the cladding holding the nuclear fuel material as disclosed in U.S. Pat. No. 3,230,150 (copper foil), German Patent Publication DAS No. 1,238,115 (titanium layer), U.S. Pat. No. 3,212,988 (sheath of zirconium, aluminum or beryllium), U.S. Pat. No. 3,018,238 (barrier of crystalline carbon between the UO.sub.2 and the zirconium cladding), and U.S. Pat. No. 3,088,893 (stainless steel foil). While the barrier concept proves promising, some of the foregoing references involve incompatible materials with either the nuclear fuel (e.g., carbon can combine with oxygen from the nuclear fuel), or the cladding, (e.g., copper and other metals can react with the cladding, altering the properties of the cladding), or the nuclear fission reaction (e.g., by acting as neutron absorbers). None of the listed references disclose solutions to the recently discovered problem of localized chemical-mechanical interactions between the nuclear fuel and the cladding.
Further approaches to the barrier concept are disclosed in U.S. Pat. No. 3,969,186, issued July 13, 1976 (refractory metal such as molybdenum, tungsten, rhenium, niobium and alloys thereof in the form of a tube or foil of single or multiple layers or a coating on the internal surface of the cladding), and U.S. Pat. No. 3,925,151, filed Feb. 11, 1974 (liner of zirconium, niobium or alloys thereof between the nuclear fuel and the cladding with a coating of a high lubricity material between the liner and the cladding).
Accordingly, it has remained desirable to develop nuclear fuel elements minimizing the problems discussed above.